Accelerating fissile material detection with a neutron source

ABSTRACT

A system discriminating fissile material from nonfissile material wherein a digital data acquisition unit collects data at high rate, and processes large volumes of data directly to count neutrons from the unknown source and detect excess grouped neutrons to identify fission. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and inducing neutron emission therefrom, and a DC power supply that exhibits electrical ripple of less than one part per million. A neutron count histogram and Poisson count distribution are overlaid to provide a visual indication of the difference in correlation of natural and induced emitted neutrons from the radiation source to characterize the neutron source as fissile material or non-fissile material.

CROSS-REFERENCE TO RELATED APPLICATIONS

This application is a third Divisional Application of U.S. patentapplication Ser. No. 14/213,502, filed Mar. 14, 2014, which is aContinuation-In-Part of U.S. patent application Ser. No. 12/712,040filed on Feb. 24, 2010 entitled “Fission Meter and Neutron DetectionUsing Poisson Distribution Comparison,” now U.S. Pat. No. 8,891,720issued on Nov. 18, 2014, which is a Continuation-In-Part of U.S. patentapplication Ser. No. 11/233,228, filed on Sep. 21, 2005 entitled“Fission Meter,” now U.S. Pat. No. 8,155,258 issued on Apr. 10, 2012,which in turn claims the benefit of U.S. Provisional Patent ApplicationNo. 60/612,968 filed by Mark S. Rowland and Neal J. Snyderman Sep. 24,2004 and titled “Fission Meter.” U.S. Provisional Patent Application No.60/612,968 is incorporated herein by this reference.

STATEMENT REGARDING FEDERALLY SPONSORED RESEARCH

The United States Government has rights in this invention pursuant toContract No. DE-AC52-07NA27344 between the United States Department ofEnergy and Lawrence Livermore National Security, LLC.

FIELD

The present invention relates to nuclear fission and more particularlyto a system for accelerating detection of fissile material.

BACKGROUND

The detection and interdiction of illicitly trafficked Special NuclearMaterial (SNM) is very important in the ongoing anti-terroristactivities undertaken by homeland security agencies. U.S. PatentApplication No. 2005/0105665 by Lee Grodzins and Peter Rothschild for asystem of detection of neutrons and sources of radioactive material,published May 19, 2005, provides the following state of technologyinformation: “There is a need to find sources of radiation and othernuclear material that are clandestinely transported across nationalboundaries. The sources of clandestine nuclear material may be in theform of “dirty bombs” (e.g., a conventional explosive combined withradioactive nuclides designed to spread radioactive contamination upondetonation), fissile material, and other neutron and radiation emittingsources that may present a hazard to the public. During recent years,the United States government has placed mobile vehicles at strategicareas with gamma ray detectors dedicated to the task of finding fissilematerial. Atomic explosives may be made from ²³⁵U, a rare, naturallyoccurring, isotope of uranium that lives almost 10⁹ years, or ²³⁹Pu, areactor-made isotope that lives more than 10⁴ years. ²³⁵U decays withthe emission of gamma ray photons (also referred to as ‘gammas’),principally at 185.6 keV and 205.3 keV. ²³⁹Pu emits a number of gammarays when it decays, the principal ones being at 375 keV and 413.7 keV.These gamma rays are unique signatures for the respective isotopes. Butfissile material invariably contains other radioactive isotopes besidesthose essential for nuclear explosives. For example, weapons gradeuranium may contain as little as 20% ²³⁵U; the rest of the uraniumconsists of other isotopes. The other uranium and plutonium isotopesreveal their presence by gamma rays emitted by their daughters. Forexample, a daughter of ²³⁸U emits a high energy gamma ray at 1,001 keV;a daughter of ²³²U, an isotope present in fissile material made in theformer USSR, emits a very penetrating gamma ray at 2,614 keV; and adaughter of ²⁴¹Pu emits gamma rays of 662.4 keV and 722.5 keV.”

U.S. Pat. No. 4,201,912 issued May 6, 1980 to Michael L. Evans et al.and assigned to the United States of America as represented by the U.S.Department of Energy, provides the following state of technologyinformation: “A device for detecting fissionable material such asuranium in low concentrations by interrogating with photoneutrons atenergy levels below 500 keV, and typically about 26 keV. Induced fastneutrons having energies above 500 keV by the interrogated fissionablematerial are detected by a liquid scintillator or recoil proportionalcounter, which is sensitive to the induced fast neutrons. Since theinduced fast neutrons are proportional to the concentration offissionable material, detection of induced fast neutrons indicatesconcentration of the fissionable material.”

U.S. Pat. No. 3,456,113 issued Jul. 15, 1969 to G. Robert Keepin andassigned to the United States of America as represented by the U.S.Atomic Energy Commission, provides the following state of technologyinformation: “An apparatus and method of detecting, identifying andquantitatively analyzing the individual isotopes in unknown mixtures offissionable materials. A neutron source irradiates the unknown mixtureand the kinetic behavior of the delayed neutron activity from the systemis analyzed with a neutron detector and time analyzer. From the knowndelayed neutron response of the individual fission species it ispossible to determine the composition of the unknown mixture. Analysisof the kinetic response may be accomplished by a simple on-line computerenabling direct readout of isotopic assay.”

Traditional neutron detectors that have been used to augment gamma-raydetection systems typically rely on “gross-counting” to detect anincreased neutron presence that may provide an indication of elevatedfission from an unknown source. Fissile material detection with passiveneutron multiplicity counters use the observation of correlated neutronsto indicate presence of fissile sources as opposed to industrial neutronsources (e.g., AmLi or AmBe). When measuring uranium, the spontaneousfission rate that is produced requires a passive measurement on theorder of one day to see a correlated fission signal. To speed thisprocess, an external neutron source or generator may be used to inducefission in the U235. Long-lived neutron sources like AmLi may inducefission and work well when the detection mechanism uses a Poissondiscrimination technique as described herein. For enhanced operatorsafety, electrically generated neutron sources, commonly available inthe form of DD (deuterium-deuterium) or DT (deuterium-tritium) sourcesare used. These generators are neutron source devices that containcompact linear accelerators and that produce neutrons by fusing isotopesof hydrogen together. The fusion reactions take place in these devicesby accelerating the deuterium, tritium, or a mixture of these twoisotopes into a metal hydride target, which also contains deuteriumand/or tritium.

These electric neutron sources are intended to be either pulsed orsteady state in neutron production. Electric neutron sources areregularly used to induce fission, where the user irradiates a sample andthen looks for delayed neutrons as a signal that fissile material ispresent. However, such an approach is relatively hazardous andinefficient because of the high neutron intensity necessary to induceenough residual delayed fission product activity. A more efficientalternative to observing the delayed neutron fraction is to look forneutrons produced while the interrogation beam is on. This is moreefficient because the induced fission rate is controlled by the fissioncross section, unlike the delay-based method which delivers a fewpercent of this induced fission. This requires the detector todistinguish between the electric source neutrons and the induced fissionneutrons, which has traditionally been impractical with present portableDD or DT electric neutron sources.

What is needed, therefore, is a method of distinguishing between thesetwo types of detected neutrons by recognizing patterns of neutronscreated and/or counted by the system.

Another disadvantage associated with present systems is that theelectric generators used in such systems are almost always made using ACrectifying, high voltage DC power supplies. This introduces theproblematic effect of electrical ripple on the DC supply that can causecorrelation of neutron product, thus introducing unwanted correlation inthe DD or DT neutron generator.

What is further needed, therefore, is a Poisson neutron source for usein in-beam interrogation systems that imposes virtually no ripple todistort the correlation of generated neutrons in a neutron detectionsystem.

SUMMARY

Embodiments of the present invention provide a neutron detection systemthat can be used to discriminate fissile material from non-fissilematerial. In general, a fissile material is one that is capable ofsustaining a chain reaction of nuclear fission. The detection systemcomprises a low cost digital data acquisition unit that collects data athigh rate and processes in real-time, large volumes of data directlyinto information that a first responder can use to discriminatepotentially radioactive materials. The detection system includes aPoisson neutron generator for in-beam interrogation of a possiblefissile neutron source. The source is designed to have a DC power supplythat exhibits electrical ripple on the order of less than one part permillion. Certain voltage multiplier circuits are used to enhance theeffective of series resistor-inductor circuits components to reduce theripple associated with traditional AC rectified, high voltage DC powersupplies.

To overcome the above-stated disadvantages of present systems,multiplicity counting objectives and counter functions have beenincorporated with certain neutron sources and counting methods todistinguish between source neutrons versus induced fission neutrons. Ingeneral, the methods described herein are based on recognizing patternsof neutrons created and/or counted by the system. The patterns includedetectable correlation that occurs on the time scales of nanoseconds ormicroseconds, which may be achieved with scintillators or gas detectors.Detection is therefore based on recognizing the count distributioncorrelation signature shape, and recognizing the rate at which thecorrelation arrives. These steps may be used in any method combinationto meet detection requirements. For example, embodiments utilize thefact that the AmLi type neutron source is distinguished by the fact thatthe neutron output rate is Poisson, which enables a multiplicity counterto distinguish fission neutrons, which are correlated.

In the following description, numerous specific details are introducedto provide a thorough understanding of, and enabling description forembodiments of the neutron detection system. One skilled in the relevantart, however, will recognize that these embodiments can be practicedwithout one or more of the specific details, or with other components,systems, etc. In other instances, well-known structures or operationsare not shown, or are not described in detail, to avoid obscuringaspects of the disclosed embodiments.

A standard approach to locating neutron sources is to use a neutrondetector to look for count rate increases above background patterns.Given the number of legitimate neutron sources used in industry,deploying standard neutron detectors will result in a large number ofalarms that will need to be resolved by more intrusive inspections.Embodiments of the present invention provide a simple way todiscriminate the commonly used neutron sources from illicit (fissile)neutron sources. This technique functions in a passive mode much like astandard portal monitor. Embodiments also provide a system forconverting the technique to an active interrogation scheme.

Embodiments of the system identify when fission is occurring byproviding an analysis of the range of simultaneous neutrons. Fission isdefined as the emission of multiple neutrons after an unstable nucleusdisintegrates. For example, Pu240 decays at a rate of about 400 fissionsper second per gram of Pu240 atoms. When the fission occurs, multipleneutrons are emitted simultaneously, with the number ranging from zeroto eight neutrons. This simultaneous neutron emission characteristic isunique to fission. Depending on the multiplication from the mass offissionable material, the knock-on fission chain time evolution caneasily last into the millisecond time scale.

Embodiments provide a method of identifying fission from an unknownsource. The method comprises the steps of counting neutrons from theunknown source and detecting excess grouped neutrons, over some timeinterval longer than cosmic detections, to identify fission in theunknown source. In one embodiment, the step of detecting excess groupedneutrons includes plotting a Poisson count distribution on top of ameasured count distribution, such that the mean count of the data is thesame as that of the Poisson curve, and discerning differences attributedto fission in the unknown source.

A fission meter apparatus for identifying fission from an unknown sourceis also described. The fission meter apparatus comprises a multiplicitycounter that looks for a range of excess neutrons from the unknownsource, a neutron detector operatively connected to the multiplicitycounter, and a calculating system operatively connected to the neutrondetector that is set up to compute a difference between actual andexpected neutron group sizes, which then positively identifies fissionin the unknown source. In one embodiment, the calculating system is asystem for plotting a Poisson count distribution superimposed over ameasured count distribution, such that the mean count of the data is thesame as that of the Poisson curve. The apparatus includes a graphingcomponent that displays the plot of the neutron distribution from theunknown source over a Poisson distribution and a plot of neutrons due tobackground or environmental sources. A known neutron source can beplaced in proximity to the unknown source to actively interrogate theunknown source in order to accentuate differences in neutron emissionfrom the unknown source from Poisson distributions and/or environmentalsources.

Although the described embodiments are susceptible to modifications andalternative forms, specific embodiments are shown by way of example, andit should be noted that the invention is not limited to the particularforms disclosed. The described embodiments cover all modifications,equivalents, and alternatives falling within the spirit and scope of theinvention as defined by the claims.

BRIEF DESCRIPTION OF THE DRAWINGS

The accompanying drawings, which are incorporated into and constitute apart of the specification, illustrate specific embodiments of theinvention and, together with the general description of the inventiongiven above, and the detailed description of the specific embodiments,serve to explain the principles of the invention.

FIG. 1 illustrates a method of detecting fission from unknown andpotentially dangerous sources of nuclear radiation, under an embodiment.

FIG. 2 illustrates a system for detecting fission from unknown andpotentially dangerous sources of nuclear radiation, under an embodiment.

FIG. 3A illustrates an example plot of the count distribution of thefrequency of neutrons emitted from an unknown source counted in adefined duration count gate.

FIG. 3B shows a plot of a Poisson count distribution on top of themeasured count distribution.

FIG. 3C shows a plot of a Poisson count distribution compared tobackground radiation.

FIG. 4 illustrates another embodiment of a system constructed inaccordance with the present invention.

FIG. 5 is a table that illustrates a background count distribution.

FIG. 6 illustrates a neutron distribution curve illustrating a cosmicsource.

FIG. 7 illustrates a neutron detection system with active interrogation,under an embodiment.

FIG. 8 illustrates a neutron detection system including a Poissonneutron electric source, under an embodiment.

FIG. 9 is a block diagram of a no-ripple DC power supply for use in aPoisson neutron electric source, under an embodiment.

INCORPORATION BY REFERENCE

Each publication, patent, and/or patent application mentioned in thisspecification are herein incorporated by reference in its entirety tothe same extent as if each individual publication and/or patentapplication was specifically and individually indicated to beincorporated by reference.

Among other references specifically cited herein, U.S. patentapplication Ser. No. 12/712,040, filed on Feb. 24, 2010, U.S. patentapplication Ser. No. 11/233,228, filed on Sep. 21, 2005, and U.S.Provisional Patent Application No. 60/612,968 filed Sep. 24, 2004 arehereby incorporated in their entirety by reference.

DETAILED DESCRIPTION OF THE INVENTION

Referring to the drawings, to the following detailed description, and toincorporated materials, detailed information about the invention isprovided including the description of specific embodiments. The detaileddescription serves to explain the principles of such embodiments, whichare susceptible to modifications and alternative forms. The describedembodiments are not limited to the particular forms disclosed. Theinvention covers all modifications, equivalents, and alternativesfalling within the spirit and scope of the invention as defined by theclaims.

FIG. 1 illustrates a method of detecting fission from unknown andpotentially dangerous sources of nuclear radiation, under an embodiment.A system implementing the method first counts neutrons emitted from thesource; block 101. It then detects grouped neutrons, block 102, andplots a Poisson count distribution on top of a measured countdistribution, block 103. Evident is the opportunity to log the timescale for the set of correlated neutrons that are counted.

An embodiment of the system 100 comprises plotting a Poisson countdistribution over a measured count distribution, such that the meancount of the data is the same as that of the Poisson curve. Thedifference between the two superimposed distributions (curves) is thenanalyzed to discern neutron emission that may be attributed solely tofission in the unknown source.

A Poisson distribution or curve is a discrete probability distributionthat expresses the probability of a number of events occurring in afixed period of time if these events occur at a known average rate andare independent of one another. The Poisson distribution formula is asfollows: f(k;λ)=(e⁻ ^(λ) λ^(k)/k!) where k is the number of occurrencesof an event and λ is a positive real number of the expected number ofoccurrences during the given interval.

The system implementing the method of FIG. 1 can be used for mobile orstationary monitoring and characterization of the type of neutronsources inside packages. Some examples of uses of the system 100 includeinspection of packed cargo containers and trucks. The present inventioncan be used for preventing illicit trafficking of fissioning nuclearmaterial, can be used for the management of inventories of nuclearmaterial, and can be used for management of waste streams of nuclearmaterial. The system 100 is particularly useful where the desire is tohave a simple, quick approach that minimally trained operators can useto improve the control of fissioning material.

In physics, fission is defined as the emission of multiple neutronsafter an unstable nucleus disintegrates. For example, Pu240 decays at arate of about 400 fissions per second per gram of Pu240 atoms. When thefission occurs, multiple neutrons are emitted simultaneously, with thenumber ranging from zero to eight neutrons. The present inventionprovides a system that can be used to identify when fission occurs bylooking for the range of simultaneous neutrons. This simultaneousneutron emission characteristic is unique to fission. Embodiments aredirected to a system that includes a multiplicity counter and a neutrondetector that is set up to observe the presence of time grouped neutronsin order to detect the simultaneous emission of neutrons.

The method and system corresponding to that illustrated in FIG. 1 hasmany uses. For example, one use of the method comprises preventingillicit trafficking of fissioning nuclear material. Another use of themethod comprises management of inventories of nuclear material. Anotheruse of the method comprises management of waste streams of nuclearmaterial. The method and system of FIG. 1 is particularly useful wherethe desire is to have a simple, quick approach that minimally trainedoperators can use to improve the control of fissioning material. Theoperators, for example may include border or traffic police, baggagehandlers or freight companies, or for international treaty agreementsthat endeavor to identify, segregate, or manage the world's inventoriesof nuclear material.

Referring now to FIG. 2, another embodiment of a system constructed inaccordance with the present invention is illustrated. This embodiment ofthe system is designated generally by the reference numeral 200. Thesystem 200 comprises a number of interconnected the structuralcomponents. A neutron detector 201 detects neutrons, a multiplicitycounter 201 looks for a range of simultaneous neutrons from the unknownsource. A calculator 203 calculates the difference between a Poissondistribution and an unknown distribution. The neutron detector 201 isoperatively connected to the multiplicity counter 202. The calculator203 is operatively connected to the multiplicity counter 202 and is setup to see time grouped neutrons to see simultaneous neutrons andidentify fission from the unknown source.

The system 200 provides a simple way to discriminate the commonly usedneutron sources from illicit (fissile) neutron sources. The system 200comprises a fission meter apparatus for identifying fission from anunknown source. The fission meter apparatus 200 comprises a multiplicitycounter 202 that looks for a range of excess neutrons from the unknownsource, a neutron detector 201 operatively connected to the multiplicitycounter, and a calculating system or analysis component 203 operativelyconnected to the multiplicity counter 202 that includes a differencecalculator 205 to compute a difference between actual and expectedneutron group sizes, which when positively identifies fission in theunknown source. In one embodiment, the analysis component 203 alsoincludes a graph display component for plotting a Poisson countdistribution and graphically displaying it as superimposed on a measuredcount distribution, such that the mean count of the data is the same asthat of the Poisson curve.

In one embodiment, the detector 201 is a neutron detector subsystem thatconsists of multiple moderated 7.5 atmosphere Helium-3 (³He) neutrondetectors. The detector subsystem includes high voltage supplies for theHelium tubes and preamplifier or discriminator units required to achievethe pick-off (timing) of the neutron events. Depending uponconfiguration, the detector may consist of two or more large(photodetectors or anode wire charge collectors) avalanche photodiodesviewing a gas volume filled with the pressurized Helium. Neutrons aredetected through scintillation and ionization of the Helium.

A wavelength shifting process, such as that known to those of ordinaryskill in the art, may be used to measure the degree of scintillation inorder to provide a measure of neutron count in the photodiodes.Alternatively, the ionized He3 and a small amount of buffer gas like CO2or Argon will ionize under the proton emission from neutron capture inHe3. In a further alternative, the scintillation from neutroninteractions in a volume of liquid scintillator common to the countingfield may substitute for the scintillation in the gaseous activedetector volume. The detector 201 gathers the neutron data and analyzesthe data for coincidences, which are doublets, triplets, quads, or anymultiplet up to a high order arriving over a logged time scale. Neutronmultiplicities in various time sub-gates (the time scale indicator)during each data acquisition cycles are recorded. An acquisition cyclemay be defined as 512 time bins. The multiplicity counter 202 maycomprise an electronic subsystem that processes the count data from thedetection system. The relative time intervals between neutrons arrivingat the detector are measured to build a statistical distribution of themultiplicity of the neutron detection.

In one embodiment, the multiplicity counter takes each detected neutronand looks in up to 512 time interval gates to record the time intervalbetween each neutron and others in the data stream from the detector.From the same data stream the shorter time intervals (e.g., on ananosecond time scale) carry information unique to cosmic spallationinduced neutrons. The time bins define counting gates that are triggeredby a trigger conditions. The trigger condition may be the detection of afirst neutron. The detection of additional neutrons after the triggerneutron and within the longer time bins constitutes a pair, or more, ofobserved neutrons.

As further shown in FIG. 2, the analysis component 203 includes adifference calculator that analyzes the output from the multiplicitycounter to determine if it is consistent with a background noise, aninnocent source, or a potentially dangerous radioactive source. Theanalysis component 203 includes a difference calculator 205, whichcalculates the difference between the unknown source and a standardPoisson distribution, and a graph display that displays the neutronemission distribution of the unknown source and the Poisson distributionin a superimposed graphical representation. In one embodiment, theanalysis component 203 performs an analysis of the neutron multiplicitydata through a Feynman Variance Technique, or equivalent method.

In one embodiment of the system 200, the analysis component 203 includesa plotting system for plotting a Poisson count distribution on top of ameasured count distribution, such that the mean count of the data is thesame as that of the Poisson curve. In an embodiment, the plotting system203 is a computer. The system 200 provides a neutron detector that canbe used to discriminate fissile material from non-fissile material. Itcomprises a low-cost digital data acquisition unit that collects data athigh rate, and in real-time processes large volumes of data directlyinto information that a first responder can use to discriminate varioustypes of materials.

Neutron Count Plots

One significant characteristic of fission is that neutrons emit ingroups. Random sources of neutrons are emitted with no regard forgrouping, however, since the appearance of these neutrons at thedetector are randomly spread in time, some may accidentally appear inclose temporal proximity. An example is a neutron detector that countsneutrons for short periods of time, for example ½ millisecond timeperiods (gate periods). This example time corresponds to a typicalneutron diffusion time in a typical moderated detector, the choice ofwhich depends on specifics related to detector design. If the ½millisecond period is counted once, the count may be one, two, or threecounts, or some other integer number, including zero. It is desirable toselect an appropriate observation time, such as two to three times thetypical neutron diffusion time, and then repeat the sampling of countsperiod many times to produce a histogram of counts described as thenumber of occurrences of each multiplet group. This yields adistribution of the number of times (e.g., 0, 1, 2, 3) that neutronswere observed over a number of detection periods (e.g., 10,000 repeatedperiods). In the case of scintillation counters, these need not bemoderated, and this may be preferred as appropriate to the countersystem design, so that the nanosecond time scales of the timecorrelations may be differentiated from the longer fission chainevolution.

FIG. 3A illustrates an example plot of the count distribution of thefrequency of neutrons from an unknown source counted in a 512microsecond count gate. For the example plot 301 of FIG. 3A, it can beseen that eight neutrons were observed 10⁵ times and 25 neutrons wereobserved about 100 times. The observed plot 301 provides an indicationof the detection of coincidental neutrons (e.g., two or more neutronsemitted within a defined time period after detection of the firstneutron) during a particular time gate.

Fission is unique in that it creates real correlations, whilenon-fission neutron sources create accidental correlations. Embodimentsprovide a method and system that utilizes new developments in howfission neutron chains are modeled to simplify and remove problemsrelated to the assay of unknown packages of fissioning material.

Counting neutrons by looking for time-correlated groupings is calledmultiplicity counting. The groupings arise from the fission processwhere a portion of a fission chain is detected. The analysis of thistype of data assists in deriving mass, multiplication, detectorefficiency, and alpha ratio (mMeA). Other factors in the analysisinclude neutron lifetime (L=1/λ), measurement gate width (T), themaximum size of neutron multiplets observed (n), the backgroundcorrelation and count rate (B), and the generalized Poisson exponent(A).

Referring now to FIG. 3B, a plot further illustrates the embodiments 100and 200. The plot is designated generally by the reference numeral 310.The top curve 301 is a count distribution of the frequency of neutronsfrom an unknown source counted in a 512 microsecond count gate, such asthat illustrated in FIG. 3A. Correction for cosmic induced correlationis performed by selecting only the correlation time scales greater thantens of nanoseconds. For example, eight neutrons were observed 10⁵ timesand 25 neutrons were observed about 100 times. The bottom curve 312 is aPoisson count distribution with the same mean count i.e., about seven.As can be seen in FIG. 3B, there is an increase in frequency of dataabove the Poisson points. That is, the actual distribution curve 301exhibits a greater number of observed neutrons above the mean count thandoes the Poisson curve 312. This represents an excess number of emittedneutrons from the unknown source over the statistically expected numberrepresented by the Poisson curve 312. If an operator observes such anexcess, either visually or via a numerical subtraction, then fission isidentified.

The actual amount of excess that triggers the detection of fission isdefined by the constraints of the system and normal operating practice.The error bars 316 represent a range of error assigned to each count. Ifthe actual number of neutrons exceeds the Poisson number but is withinthe error range, then fission may not be cause of such excess. However,if the actual number of neutrons exceeds the error range of the Poissoncount by a pre-defined amount, then such an excess may be attributed tofission.

In general, the presence of background radiation (e.g., cosmic rays) maybe a factor in any detection process. However, methods of the fissionmeter plot described herein are still useful and generally notoverwhelmed by background effects. In certain cases, a very weak fissionsource may be overwhelmed by combinations of background noise, however,a fission source that is practically detectable will have a countdistribution curve that is similar to the Poisson distribution, as shownin FIG. 3B. Embodiments include a method for distinguishing backgroundradiation to further refine the detection of fissioning material.Background radiation may be correlated to some degree, but has a verydistinct count distribution curve. It has a flattened out portion aftera certain number of counts, and does not monotonically decrease, as doesa Poisson distribution.

FIG. 3C shows a plot of a Poisson count distribution compared tobackground radiation. As shown in FIG. 3C, a count distribution forobserved background data 324 is plotted relative to a Poissondistribution 322. A pure background source will show a curve thatflattens or has a kink shape around counts 3 or 4, as shown in FIG. 3C.Therefore, a detectable radiation source will have a count distributionthat resembles the Poisson shape, but with no kink, and depending on itsstrength, it will overwhelm the background effects in the 3 and 4 countregion. The practical range of filtering out background depends onvarious parameters associated in specifying a neutron detector, such asefficiency, distance from source, and so on. In a typical application,background count rates may be on the order of 3.5 counts/second (cps). ACf (Californium) fission source with Multiplication=1 typically makesone million neutrons per second; at a distance of one meter, thedetector efficiency is around 1% so the count rate would be thousands ofcps. Such an example overwhelms the background effects. For asignificant amount of fissioning material (e.g., tens of kilograms ofuranium), for which the Multiplication=10, at one meter the count rateis 3 cps so the total count rate would be 6.5 cps. There is a cleardeviation of 3, 4, 5 counts because of the multiplication, and thehigher multiplets overwhelm background even though the count rate isnear background. Through the graph display process 206, the generatedcount distribution plots will show that there is no “flat” portion onthe tail on the observed count plots, unlike the background data shownin FIG. 3C. Thus, this method provides a means of distinguishing truefissioning sources from mere background and provides a basis forcomparing a non-background source with a Poisson distribution. For casesin which the detector is within range of a signal from a fissioningsource, it will report a distinction from both Poisson and backgroundcorrelation.

While cosmic induced spallation neutrons can be counted and appear ascorrelated, they occur in a time scale that can be simply segregatedform the longer fission chains of interest. In the case of a spallationneutron causing a legitimate fission chain in the hypothetical mass ofnuclear material, this will generate a legitimate fission chain worthyof detection since it is a valid signature of the desired object.

The method and systems 100 and 200 comprise a first step of countingneutrons from the unknown source and a second step of detecting excessgrouped neutrons to identify fission in the unknown source. In anotherembodiment the method and systems 100 and 200 comprise the steps ofcounting neutrons from the unknown source and detecting excess groupedneutrons to identify fission in the unknown source wherein said step ofdetecting excess grouped neutrons to identify fission in the unknownsource includes plotting a Poisson count distribution on top of ameasured count distribution, such that the mean count of the data is thesame as that of the Poisson curve, and discerning differences attributedto fission in the unknown source. In another embodiment the method andsystem 100 and 200 comprise the steps of counting neutrons from theunknown source and detecting excess grouped neutrons to identify fissionin the unknown source includes plotting a Poisson count distribution ontop of a measured count distribution, such that the mean count of thedata is the same as that of the Poisson curve, and discerningdifferences attributed to fission in the unknown source and wherein saidstep of discerning differences attributed to fission in the unknownsource comprises discerning visible differences in said Poisson countdistribution superimposed over a measured count distribution plot thatare attributed to fission in the unknown source.

The process illustrated in FIG. 3B of plotting the multiplet structureof the Poisson distribution and comparing it in a multiplet-by-multipletfashion with the observed distribution is unbiased by any expectationthat the triggering event (e.g., a trigger neutron) is the correctneutron with regard to whether it is a real or accidental. Thisautomatically alerts the user to the correct and exact expected rate ofaccidental multiplets greater than one, and prevents the problem causedby systems that assume that all counts within the A gate areaccidentals, which leads to the possible rejection of valid correlationinformation.

In general, neutrons are used in many industrial applications. Neutronsignatures also indicate the presence of fissioning nuclear material. Itis desirable to be able to separate benign industrial neutron sourcesfrom fission sources. Traditionally, detection of nuclear material hasbeen accomplished by neutron counting. If neutron sources were rare, themisinterpretation of any neutron source as a fission source would be oflittle consequence. However, with the large-scale introduction ofnuclear monitoring equipment in daily commerce comes the need to notconfuse the traffic of industrial sources with illicit traffic. Themethod and systems 100 and 200 provide the basis for a visual orautomated comparison of raw count distribution data, to a Poissondistribution with the same mean count, to show graphically the intuitivesense that the characteristic of fission is present. Optimally, theexcess correlation, above a Poisson rate of correlation, may bealternatively or additionally provided by observing that numericalcharacteristics of the data and the corresponding Poisson distributionmay be computed to form a numerical difference, redundantly indicativeof fission.

As stated above, the characteristic of fission is that neutrons emit ingroups. That is, potentially dangerous unknown sources emit multiplecoincident neutrons. This simultaneous emission is used in the detectorsystem described herein to distinguish from random sources of neutronsthat are emitted with no regard for grouping; however, since theappearance of these neutrons at the detector are randomly spread intime, some may accidentally appear in close temporal proximity. Fissionis unique in that it creates real correlations, while non-fissionneutron sources create accidental correlations. Unrecognized is therelative histogram comparison of the measured or unknown neutron source,with a mathematically generated count histogram that represents thehypothetical case of no fission. Visually, in isolation, one histogramlooks like another. FIG. 3B illustrates a detector system that includesa histogram display system that allows direct graphical comparison ofthe measured source to the mathematically generated or Poissondistribution. The shape of the measured source histogram is derived fromthe characteristics of the measured unknown source. For the example ofFIG. 3B, the tail portion of the histogram 301 is above the random orPoisson histogram 312. This excess correlation is due to fission,illustrating that a simple plot of data collected in one measurement,can be analyzed with a relatively simple procedure involvingstraightforward observation and comparison. Alternatively, it ispossible to compute various quantities in order to derive mathematicalcount differences between the histograms in order to obtain numericmeasures of excessive neutron emission. Threshold values can be definedsuch that automated processes can indicate the presence of a potentiallydangerous source if the difference between the measured count exceedsthe Poisson count in excess of the threshold.

One example of an alternative embodiment to the histogram overplotconcept is to numerically compute quantities based on the singlemeasurement of an unknown. Conceptually, the objective is to realizethat the differences apparent in a comparison of histograms may bedescribed as the number of pairs of counts observed in the unknown minusthe number of count pairs expected if there were no fission (but theneutrons came from a non-fissioning neutron source). This can beexpressed as:# of pairs observed−expected random # of pairs

If the difference is zero, then the observed neutron source is notundergoing neutron fission. The number of pairs is only one example of astatistical quantity derivable from the measured histogram. Others mightbe the third or fourth moment of the histogram.

An alternative embodiment to the graphical histogram approach involvesan analysis of the number of pairs of neutrons. As stated above, pairsof neutrons in excess of those expected is the test. Numerically thismay be computed from the measured histogram:

${\sum\limits_{n = 0}^{\infty}\;{\frac{\frac{n\left( {n - 1} \right)}{2}}{2{\sum\limits_{n = 0}^{\infty}\;{Cn}}}{Cn}}} - {\left( \frac{\sum\limits_{n = 0}^{\infty}\;{nCn}}{\sum\limits_{n = 0}^{\infty}\;{Cn}} \right)^{2} \cdot \frac{1}{2}}$

This difference represents the absolute number of pairs in excess ofthat expected from a non-fissioning neutron source. In the aboveequation, n is the x-axis of the histogram and is the size of the groupof neutrons observed, and Cn is the number of times that a group of nneutrons was observed after repeating the ½ msec. measurement a largenumber of times. Note that the mean count of the measured histogramdefines the histogram of the expected or hypothetical non-fissionhistogram. The mean count of the measurement is:

${c\text{-}{bar}} = {\overset{\_}{c} = \frac{\sum\limits_{n = 0}^{\infty}\;{nCn}}{\sum\limits_{n = 0}^{\infty}\;{Cn}}}$

The histogram expected from a non-fission source will have the sameC-bar, however the shape of the histogram will be described by:

${{Cn}\text{-}{poisson}} = {\frac{{\overset{\_}{c}}^{n}}{n!}e^{- \overset{\_}{c}}}$

In the above equation, n is the count group size. Whether the systemsimply plots Cn-Poisson on top of the measurement, as in the firstembodiment, or computes difference quantities, as in the secondembodiment, they represent the same insight that a uniquely observablefission neutron signature can be created from a single measurement, andcan be useable by minimally trained operators to separate high valueobjects from common industrial sources.

Referring now to FIG. 4, another embodiment of a system constructed inaccordance with the present invention is illustrated. This embodiment ofthe system is designated generally by the reference numeral 400. Theneutron detector 401 detects neutrons. The neutron detector 401 isoperatively connected to a counter 403. The arrow 402 illustrates pulsessent from the detector 401 to the counter 403. Pulses are sent to thecounter 403 when neutrons are captured.

The counter 402 and is set up to see time grouped neutrons to seesimultaneous neutrons and identify fission from the unknown source. Thecounter 402 (1) can record how many counts (group size) arrive in a ½millisecond period, (2) repeat the ½ millisecond recording period manytimes, and (3) plot a histogram of the number of times the differentgroup sizes occur.

The counter 403 is operatively connected to a plotter or differencecalculator 405. The arrow 402 illustrates information from the counter403 being sent the plotter or difference calculator 405. The system 400provides a simple way to discriminate the commonly used neutron sourcesfrom illicit (fissile) neutron sources. In one embodiment, a systemplots a Poisson count distribution on top of a measured countdistribution, such that the mean count of the data is the same as thatof the Poisson curve. Such a comparison plot is shown in FIG. 3B.

In one embodiment, the neutron detector is used in a portable neutronsource identification system that helps detect the presence of illicitradioactive material for use in homeland security applications. Suchmaterial can be used in deadly terrorist weapons such as ImprovisedNuclear Devices (IND) or state built nuclear weapons. In general, theseweapons require the presence of a so-called Special Nuclear Material(SNM), that is, Uranium or Plutonium, to create a nuclear explosion.Traditional methods of detecting and identifying the presence of SNMinvolve the use of gamma-ray detection. These methods, however, can bedefeated through the use of heavy metal shielding. The neutron detectoraccording to embodiments augments the technique of gamma-ray detectionby identifying fission neutron sources by examining the inherentcharacteristics of the neutron decay process. The neutron detector underembodiments includes processing and filtering components that not onlycount neutrons, but also check the source and environmental conditionsfor the existence of neutron sources beyond simple noise orenvironmental effects. Such a detector allows for the rapid andrelatively certain detection of neutron sources from potentiallydangerous sources, such as improvised nuclear devices or similarweapons.

A neutron source can be any of a variety devices that emit neutrons,irrespective of the mechanism used to produce the neutrons. Dependingupon variables including the energy of the neutrons emitted by thesource, the rate of neutrons emitted by the source, the size of thesource, neutron source devices can be found in a diverse array ofapplications in areas of physics, engineering, medicine, nuclearweapons, petroleum exploration, biology, chemistry, nuclear power andother industries. Man-made sources include reactors that produceneutrons that can be used for experiments, and spallation sources thatare high flux sources, in which protons that have been accelerated tohigh energies hit a target material, prompting the emission of neutrons.

In one embodiment, the neutron detection system includes a method forallowing the filtering of background neutron noise due to other sources,such as cosmic or man-made sources. Typical background consists ofsingle neutrons and neutron groups from multiple neutron events causedby cosmic rays. The Poisson distribution of the events will cause somerandom coincidence events. These random coincidences can be calculatedusing the singles count rate and device characteristics. FIG. 5 is atable that illustrates a background count distribution for an exampletime period. For table 500 of FIG. 5, data was collected for a period ofone hour resulting in a count of 8552 for a count rate of 2.31 countsper second (cps).

FIG. 6 illustrates a neutron distribution curve illustrating a cosmicsource. In one embodiment, the simple observation of a neutrondistribution curve with a shape like that shown in FIG. 3A wouldindicate the presence of neutrons due to cosmic interference.Correlation is indicated by the presence of events with higher ordermultiplicity in the distribution. As shown in FIG. 6 the actualbackground 603 is slightly more correlated than the neutron distributionfrom the unknown source 601, and both are more correlated than the purePoisson distribution 602. Such an effect is also shown in FIG. 3C. Asshown in FIG. 6, the actual background curve 603 has a characteristicand relatively pronounced curve up at the very end of the plot. Theshape of curve 603 can be used by an analyst or a program to determinewhether or not the presence of neutron emission is due to cosmic effectsas opposed to a potentially dangerous source.

The distribution curves 601, 602, and 603 shown in FIG. 6 provide agraphical basis on which an analyst can view and identify man-made orenvironmental sources of neutrons. The difference in counts above themean, that is, in the upper portion of each curve, along with the shapeof the curve can be used to characterize the criticality of the hazardposed by an unknown source relative to the background and Poissondistributions. In one embodiment, analysis of the graphical neutrondistribution data as generated by the neutron detection system can beviewed and analyzed by a human operator.

Alternatively, the graphical distribution data can be further processedin a program or electronic module to provide an interpretation of thedata. This module can be configured to analyze one or more parametersassociated with the distribution plot such as shape, rate of rise of aportion of the curve, point-by-point differences with the Poisson and/orenvironmental neutron plots, and so on. Such interpretation informationcan be used by a user or a further response system to trigger anappropriate response to the unknown source, such as sounding an alarm,ordering an evacuation, initiating an automatic detonation sequence, orany other appropriate action.

In one embodiment, the detection system includes a module that allowsfor active interrogation of an unknown neutron source. This systemincludes a source of neutrons, such as Californium (Cf) orAmericium-Beryllium (AmBe) placed at a known distance from the unknownsource. The active interrogation due to the presence of a neutron sourceeffectively forces neutrons into the source and results in morefissions. This generally increases the speed in which the neutrondistribution for the unknown source is generated. The resulting neutrondistribution is then observed. FIG. 7 illustrates a neutron detectionsystem with active interrogation, under an embodiment. In system 700,unknown source 704 is placed in the proximity of detector 706. Thedetector 706 also picks up neutron emissions from background source 702.To counteract the effects of this background noise, a known source 708is used to drive neutrons into the unknown source 704. The resultingneutron emission distribution is then plotted relative to a Poissondistribution, and a graph, such as that shown in FIG. 3 is displayedusing graph generator 710. The active interrogation system of FIG. 7 canincrease the strength of the unknown source above the ten to one ratiorelative to the background, thus allowing greater possibility ofdetection from unnatural sources.

Accelerating—Fissile Material Detection

As described above, fissile material detection with a passive neutronmultiplicity counter is relatively well-known, and the basic detectionmechanism depends on the observation of correlated neutrons, where thecounter will clearly show that fission creates correlated neutrons incontrast to an industrial neutron source (like AmLi or AmBe) that doesnot create correlated neutrons.

FIG. 3B illustrates the comparison of plots for an uncorrelated(Poisson) neutron source 312 with a correlated (observed) neutron source301. The attribute that describes the uncorrelated data is that of atime-random neutron source, and is described by the Poisson probabilitydistribution: b_(n)=(C^(n)/n!)e^(−C). In this equation, b is theprobability of detecting a particular number of neutrons, n, during acounting window (e.g., 512 microseconds), C is the average number ofcounts observed during the counting window. In the plots of FIG. 3B, theobserved curve 301 is above Poisson curve 312 because the neutronsdetected were created by a fission process, where fission neutrons arecreated simultaneously in groups ranging in size from zero to abouteight neutrons. The groups of neutrons from fission are therefore alsodetected with groupings that favor the higher multiplets, as shown inFIG. 3B.

Neutron generators generally contain compact linear accelerators thatproduce neutrons by fusing isotopes of hydrogen together. A DD neutronsource fuses deuterium atoms (D+D) to form an He-3 ion and a neutronwith a kinetic energy of about 2.5 MeV. A DT neutron source fuses adeuterium and a tritium atom (D+T) to form an He-4 ion and a neutronwith a kinetic energy of about 14.1 MeV. Neutrons are produced byaccelerating Deuterium and/or Tritium ions into a hydride target loadedwith Deuterium and/or Tritium. Deuterium atoms in the beam fuse with theDIT atoms in the target to produce neutrons. Neutrons produced from theDT reaction are emitted uniformly in all directions from the target,i.e., isotropically, while neutrons from the DD reaction are slightlypeaked in the forward direction. In both cases, the associated He nuclei(alpha particles) are emitted in the opposite direction of the neutron.

When measuring uranium, the spontaneous fission rate produces about 13neutrons per second per kilogram of U238 and less for U235. At thiscount rate, it might take on the order of day of counting to see thecorrelated fission signal from uranium. To speed this process, systemsmay use an external neutron source to induce fission in the U235. Usingan isotopic neutron source means the user must carry radioactivematerial, which is undesirable. An alternative approach is to use anelectrically generated neutron source, commonly available in the form ofDD or DT electric neutron sources.

In a further alternative embodiment, a 60 keV neutron source may be usedto operate below the U238 fission threshold. In general, the fact thatthe energy is above the fission threshold is not an issue becauseinducing fission is the objective. One simply needs to keep track ofwhich sources are counted in the Poisson or fission category. Withregard to the alternative embodiment, the 60 KeV neutron source may beconfigured as described in U.S. patent application Ser. No. 12/976,216,which is assigned to the assignee of the present application, and whichis incorporated herein by reference in its entirety.

The electric neutron sources are intended to be either pulsed or steadystate in neutron production. Electric neutron sources are regularly usedto induce fission, where the user irradiates a sample, and looks fordelayed neutrons as a signal that fissile material is present. Thedelayed neutrons created amount to about 2% of the fissions induced bythe electric neutron source. This method requires injecting such a largenumber of neutrons that the operation is hazardous to humans. Typicalneutron interrogators supply 10⁸ to 10¹⁰ neutrons per second and supplyneutrons for about ten minutes. The neutron generator is then turned offand the neutron counter counts any delayed neutrons to indicate fissilematerial is present.

An alternative to observing the delayed neutron fraction is to look forneutrons produced while the interrogation beam is on. This improvessystem efficiency significantly (such as up to about 5000%) since theduty factor is about 100%, in this type of system. Directly observingthe promptly induced fission neutrons requires that the neutron detectorbe able to distinguish the electric source neutrons and the inducedfission neutrons, which until now has been impractical with the veryportable DD or DT electric neutron sources. An example of an impracticalenergy selection method is a time-of-flight energy spectrometer, whichobtains very high energy selectivity by dispersing the two sources ofneutrons over a 100 meter long path. This energy selectivity approach isstill not ideal since both the electric neutron sources and inducedfission make overlapping neutron energies, making a clean separationimpossible. Energy resolving neutron spectrometers, in the form ofhand-held detectors, resolve neutron energies so inefficiently that theyare not used with the common DD and DT electric neutron generators.

To overcome this disadvantage, embodiments are directed to a superiormethod to detect fissile or fertile nuclear material using a novelneutron source that facilitates distinguishing of the two sources ofneutrons using a multiplicity counter. This method enables the use ofthe common DD or DT neutron generators. A special feature of theinterrogator is the random or Poisson time correlation of the neutronsproduced by the DD or DT electric source, coupled with a multiplicitycounter to observe any correlation. Any correlation would mean fissionwas induced and therefore fissile or fertile material is present.

In general, DD and DT electric neutron sources actually produce someamount of correlated neutrons because their electric drive circuits areactually time-varying, imprinting the high voltage ripple on the neutronoutput. Ripple is small unwanted residual periodic variation of the DCoutput of a power supply which has been derived from an AC source, suchas due to incomplete suppression of the alternative waveform in thesupply. This problem has forced all prior neutron interrogation methodsto induce fission, then turn off the neutron source, after which theuser measures the delayed neutron activity, which only comes as a resultof inducing fission. The pulsed neutron sources produce correlatedneutrons, by definition, since they are pulsed with the AC ripple addingcorrelation. Therefore, the “steady state” versions of the DD and DTsources have been, in reality, not steady state, and therefore notPoisson.

In general, Poisson DD or DT generators are always made with an ACrectifying high voltage DC supply. As stated above, the problem with allDD and DT generators is the presence of ripple on the DC supply. Thisripple can significantly correlate the neutron production, defeating theexpectation there will be no detected correlation in the absence offissile material. It is the significant amount of correlation in DD andDT neutron generators that requires all commercial waste assay countersto count only the delayed fraction of induced neutrons, which is aninefficient process, as described above. One approach to make a PoissonDD or DT neutron generator is to simply filter the AC component from thehigh voltage DC. However, this filtering technique does not work in aportable instrument because the components that would operate at therequired 120,000 volts would be larger than someone could carry.

To address this disadvantage, embodiments are directed to a steady stateneutron generator coupled with a multiplicity counter that incorporatesa circuit that strongly filters the AC from the high voltage DC signal,along with an algorithm that indicates the presence of neutroncorrelation and provides a corresponding alarm system. This system isembodied in a Poisson electric neutron source that observes any inducedneutron correlation as an indication of fission, thereby detectingfissile or fertile material. In an embodiment, such as system may bepackaged in a device that weighs less than 100 pounds. FIG. 8illustrates a neutron detection system including a Poisson neutronelectric source, under an embodiment. As shown in system 800 of FIG. 8,a neutron generator 802 of either a DD or DT electric neutron sourcetype is used to irradiate the unknown source 804 that is to becharacterized. The neutron generator is meant to induce fission in theunknown source. The detector 806 detects neutrons produced by theneutron source in a direct rather than delayed manner. The detectordistinguishes between the electric source neutrons from generator 802and the induced neutrons from source 804. The detector 806 is coupled toanalyzer 808, which includes a multiplicity counter function to observeany correlation in the detected neutrons. In an embodiment, themultiplicity counts are processed in accordance with methods describedwith respect to FIGS. 3A to 3C earlier. To eliminate the effect ofelectrical ripple or other distortion effects caused by power supplycircuitry, the neutron generator is powered by a no-ripple power supply801.

In an embodiment, the detector 806 of FIG. 8 may include two or moredifferent detector circuits to detect and distinguish neutrons ofdifferent sources. For example, background neutrons from alpha-nreactions are Poisson and have a unique count distribution shape, cosmicinduced neutrons are created in the nanosecond scale and have their ownunique count distribution shape, and fission neutrons are created over atime-scale controlled by the relatively slow fission neutron speed,which guarantees fission signatures that form in greater than tens ofnanosecond time scales up to milliseconds. To take advantage of thesedistinctions, the detector 806 may comprise at least two types ofdetectors. In an embodiment, detector 806 includes a scintillator withan energy selector that guarantees only fast and direct neutrons arecounted and a moderated neutron capture detector (He3, Boron, Lithium)for the slower times scales. Either detector may be embodied as aseparate detector within system 800 or part of an integrated or combineddetector circuit 806. In certain cases, the scintillator may also beused for the slower time scales. Other types of detectors thatdistinguish neutrons generated on a short time scale (e.g., nanosecond)versus longer time scales (e.g., milliseconds) can also be used.

As shown in FIG. 8 embodiments include a Poisson neutron source thatfeatures a characteristic for the DC supply of having a ripple less thanone part per million. At present for applications requiring 120,000volts, there are no such DC supplies. The process for filtering the DCcannot be lumped RLC (resistor-inductor-capacitor) circuits since theirvoltage stand-off and power dissipation needs would require a dielectricchamber of several feet in dimension, exceeding the volume of the normal120,000 volt supply. The Poisson neutron source under an embodiment, hastwo power supplies that have their outputs summed, with the sinusoidalripple added 180 degrees out of phase. A Cockroft-Walton voltagemultiplier is also used. The frequency of the Cockroft-Walton voltagemultiplier is increased to enhance the effectiveness of series RLcircuit components. Such a circuit significantly reduces ripple on theoutput DC signal.

FIG. 9 is a block diagram of a no-ripple DC power supply for use in aPoisson neutron electric source, under an embodiment. As shown indiagram 900, DC power supply 902 receives AC electricity in and passesthis AC input to two separate rectifier stages 904 and 906. Therectifier circuits convert the AC into DC electricity through knownrectification methods. In most practical circuits, at least some amountof ripple will be present on the DC signals that are output from therectifiers 904 and 906. The output stages of the rectifiers are coupledtogether such that the ripple components are added together 180 degreesout of phase. This effectively cancels the ripple present on the DCoutputs of both the rectifiers when the signals are added together. Thesummed DC signal is then input to a Cockroft-Walton voltage multiplier908 to bring the level of the DC output signal up to a desired level.

In general, a Cockroft-Walton voltage multiplier (or generator) is acircuit that generates a high DC voltage from a low voltage AC orpulsing DC input. It comprises a voltage multiplier ladder network ofcapacitors and diodes to generate high voltages. In an alternativeembodiment, other similar voltage multiplier circuits may be used.

The neutron detection system under an embodiment includes a Poissonneutron source, coupled with the utility of an instrument that canobserve that such a source is or is not Poisson, coupled with an alarmthat is sensitive to the distinction between Poisson and correlatedneutrons. Such a system is implemented to operate in an in-beam mode(prompt fission neutrons) that is much more efficient than previousdelay-based systems that require turning off the neutron interrogatorprior to neutron detection. A side benefit of this systematic approachis to use the large increase in total efficiency to reduce theinterrogation source strength by about a factor of 100, making portablefield use much safer.

The use of a Poisson neutron source for use in in-beam interrogationsystems that imposes virtually no ripple to distort the correlation ofgenerated neutrons in a neutron detection system reduces or eliminatesthe problematic effect of electrical ripple on the DC supply that cancause correlation of neutron product.

Time Scales and Scintillators

In an embodiment, the detector system includes a multiplicity counterfunction that uses the inherent time scales that define the correlationrate that neutron counts arrive at the detector. Thus, as describedabove, different time-scale detectors may be used in the detection andanalysis system 800 of FIG. 8. Given the timing nature inherent in thedefinition of gate width, one may simply categorize and distinguish thetime scales over which the correlated signals arrive. For example, thetime scale for cosmic spallation induced neutrons is on the order of afew nanoseconds as compared to fission chains, which must evolve over atleast tens of nanoseconds. With a gate setting of three nanoseconds anycorrelation must be from cosmic background and may be used as thedefinition of the count rate for this background and may therefore beused as a measure of how much count rate must be coming fromnon-background sources. Similarly, the alpha-induced Poisson backgroundfrom either the soil or an intentionally added interrogation source maybe removed from the total signature by the methods described inreferences such as U.S. Pat. Nos. 8,194,813 and 8,194,814, which bothare assigned to the assignee of the present application and are herebyincorporated by reference in their entirety. This leaves the user with anumerical excess in count rate and a multiplicity distribution thatdefines the slowly evolving fission signature.

As taught in the above-referenced U.S. Pat. Nos. 8,194,813 and8,194,814, correlation in count distributions may be converted to alinear space within which count rates may be subtracted, or sums ofbig-Lambda distributions may be fit in an error minimization scheme.Beyond this novelty in how to add and subtract, is the next criticalbreakthrough that allows one to directly benefit from the time scaleswithin which any correlation arrives. As an example in the distinction,a scintillator responding to neutrons reports a count to the electronicscounter in about two nanoseconds. Instead of only looking at the countdistributions that form in the microsecond to millisecond time scales,there is an opportunity to look at the nanosecond time scales. Thebenefit is that cosmic spallation produced neutron clusters are formedin time scales of nanoseconds (i.e., 3 ns). With a neutron detector likethe scintillator, this fraction of neutron counts forming correlationmay be directly and distinctively quantified, using the herein describedlinear space methods. For the purposes of detecting fissionablematerial, simply quantifying the Poisson portion of the detected signal,the cosmic induced portion and removing them from the total leaves thenet total as the basis in the method to define detection of fissionablematerial. The novel combination of methods is most generally based onrecognition of the time scale over which the correlation arrives andrecognizing the shape of this correlation in the form of the countdistribution.

As stated previously, neutrons from different sources have differentsignatures with respect to the time-scale and distribution of emission.Specifically, background neutrons from alpha-n reactions exhibit aPoisson distribution, cosmic induced neutrons are created in thenanosecond scale, and fission neutrons are created over a time scale onthe order of greater than tens of nanoseconds to milliseconds. One ormore appropriate detector circuits are provided to detect the emittedneutrons relative to these different time scales and/or emissionpatterns. The system provides combinations or subset selection of aknown signature neutron source, selection of the necessary correlationtime scales, distribution shapes, and detector time-responsespecification to enable mathematically precise access to the signaturesrequired in a particular detection system.

In one embodiment, the neutron detector system described herein can beembodied within a portable device that can be deployed in the field andused by personnel to detect the presence of potentially dangeroussources of radioactive material from virtually any type of object oritem. The packaging around any such source can be shielded orunshielded. Such a detector system can also be used in any type ofNuclear Instrumentation Module (NIM) for use in experimental particle ornuclear physics.

Embodiments of the present invention are suitable to provide a simple,quick approach that minimally trained operators can use to improve thecontrol of fissioning material. The operators, for example may includeborder or traffic police, baggage handlers or freight companies, or forinternational treaty agreements that endeavor to identify, segregate, ormanage inventories of nuclear material.

Aspects of the circuitry and methodology may be implemented asfunctionality programmed into any of a variety of circuitry, includingprogrammable logic devices (“PLDs”), such as field programmable gatearrays (“FPGAs”), programmable array logic (“PAL”) devices, electricallyprogrammable logic and memory devices and standard cell-based devices,as well as application specific integrated circuits. Some otherpossibilities for implementing aspects include: microcontrollers withmemory (such as EEPROM), embedded microprocessors, firmware, software,etc. Furthermore, aspects of the memory test process may be embodied inmicroprocessors having software-based circuit emulation, discrete logic(sequential and combinatorial), custom devices, fuzzy (neural) logic,quantum devices, and hybrids of any of the above device types. As isunderstood in the art of electronic circuit manufacture, a number ofdifferent underlying device technologies may be provided in a variety ofcomponent types.

It should also be noted that the various functions disclosed herein maybe described using any number of combinations of hardware, firmware,and/or as data and/or instructions embodied in various machine-readableor computer readable media, in terms of their behavioral, registertransfer, logic component, and/or other characteristics.

Unless the context clearly requires otherwise, throughout thedescription and the claims, the words “comprise,” “comprising,” and thelike are to be construed in an inclusive sense as opposed to anexclusive or exhaustive sense; that is to say, in a sense of “including,but not limited to.” Words using the singular or plural number alsoinclude the plural or singular number respectively. Additionally, thewords “herein,” “hereunder,” “above,” “below,” and words of similarimport refer to this application as a whole and not to any particularportions of this application. When the word “or” is used in reference toa list of two or more items, that word covers all of the followinginterpretations of the word: any of the items in the list, all of theitems in the list and any combination of the items in the list. Whileembodiments may be susceptible to various modifications and alternativeforms, specific embodiments have been shown by way of example in thedrawings and have been described in detail herein. However, it should beunderstood that the invention is not intended to be limited to theparticular forms disclosed. Rather, the invention is to cover allmodifications, equivalents, and alternatives falling within the spiritand scope of the invention as defined by the following appended claims.

What is claimed is:
 1. A system for characterizing a radiation source asfissile material or non-fissile material, comprising: a DC power supplycomprising an AC filter circuit; a neutron generator coupled to the DCpower supply and configured to irradiate the radiation source byinducing radiation in the radiation source; a detector configured toreceive and count neutrons emitted from the radiation source bothnaturally and from irradiation by the neutron generator; and an analyzercomponent coupled to the detector and configured to measure a number ofneutrons simultaneously emitted from the radiation source during anumber of measurement time periods to derive a multiplet countdistribution, compute the mean count rate of the multiplet countdistribution, compute the number of pairs of the multiplet countdistribution, use the mean count rate to produce a Poisson distributionfor the mean count rate, compute the expected number of pairs for thePoisson distribution, subtract the expected number of pairs for thePoisson distribution from the number of pairs in the measurement, andcharacterize the radiation source as fissile material if the number ofpairs in the measurement exceeds the number of pairs for the Poissondistribution.
 2. The system of claim 1 wherein the AC filter circuitcomprises: an AC receiving circuit; a pair of rectifiers converting aninput AC signal to respective DC output signals; a summer circuitreceiving the respective DC output signals and configured to add therespective DC output signals together by adding electrical ripplepresent on a DC output signal from one rectifier of the pair ofrectifiers 180 degrees out-of-phase to the electrical ripple present ona DC output signal from the second rectifier of the pair of rectifiers;and a Cockroft-Walton voltage multiplier circuit coupled to the summercircuit and configured to increase the output of the summed DC outputsignals to a desired DC voltage level.
 3. The system of claim 2 whereinthe analyzer component is further configured to: produce a hypotheticalhistogram for the Poisson distribution; produce a count histogram forthe count measurement representing a number of times different groupsizes occur from the number of measurement time periods; and overlay thehistogram and Poisson count distribution to make a mean count of thehypothetical histogram the same as for the count histogram and toprovide a visual indication of the difference in correlation of emittedneutrons from the radiation source, to characterize the neutron sourceas fissile material or non-fissile material.
 4. The system of claim 3further comprising a component to determine the value of a coefficient,Cn, which represents a number of times that a group of neutrons isobserved after repeating a ½ millisecond measurement time period adefined number of times, and wherein the mean count is calculated by theformula:$\overset{\_}{C} = {\frac{\sum\limits_{n = 0}^{\infty}\;{nCn}}{\sum\limits_{n = 0}^{\infty}\;{Cn}}.}$5. The system of claim 4 wherein the shape of the histogramrepresentation of the count measurement is given by the formula:${{Cn}\text{-}{poisson}} = {\frac{{\overset{\_}{c}}^{n}}{n!}{e^{- \overset{\_}{c}}.}}$6. The system of claim 1 wherein the neutron generator comprises apulsed electric source of neutrons irradiating the radiation sourceusing an interrogation beam of neutrons.
 7. The system of claim 1wherein the detector comprises: a first detector calibrated to a fasttime scale on the order of nanoseconds to generate a first set ofdetected neutrons; and a second detector calibrated to a slow time scaleon the order of milliseconds to generate a second set of detectedneutrons.
 8. The system of claim 7 wherein the analyzer is configured toanalyze the first and second sets of detected neutrons to determine thenumber of times that a group of n simultaneously emitted neutrons isobserved from the radiation source after a defined measurement period isrepeated a defined number of times to derive a neutron count measurementbased on at least one of the fast time scale and slow time scale todetermine whether the material is fissile versus non-fissile.
 9. Thesystem of claim 8 wherein the defined measurement time period is ½millisecond, and wherein a fissile source creates real correlationsbetween the neutrons emitted from the radiation source, and anon-fissile source creates no correlation or only accidentalcorrelations between the neutrons emitted from the radiation source. 10.The system of claim 7 wherein the first detector comprises ascintillator and energy selector system that is configured to detectfast and direct neutrons emitted from the material.
 11. The system ofclaim 10 wherein the second detector comprises at least one of amoderated neutron capture detector or a scintillator-based detector,both configured to detect neutrons emitted from the material incorrespondence with the slow time scale.